IR 05000263/1997014

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Insp Rept 50-263/97-14 on 970830-1013.Violations Noted. Major Areas Inspected:Operations,Maintenance,Engineering & Plant Support.Rept Also Includes Insp follow-up of Security Related Issue by Regional Inspector
ML20199E591
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 11/13/1997
From:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20199E574 List:
References
50-263-97-14, NUDOCS 9711210256
Download: ML20199E591 (17)
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U.S. NUCLEAR REOULATORY COMMISSIDN REGION lli

l Docket No: 50-263  !

License No: DPR 22 +

Report No: 50 263/97014(DRP)

Licensee: Northern States Power Company

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Facility: Monticello Nuclear Generating Station

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Location: 414 Nicollet Mall  ;

Minneapolis, MN 55401  ;

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Dates: August 30 October 13,1997

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Inspectors: A. M. Stone, Senior Resident inspector  !

D. Wrona, Resident inspector S. Ray, Senior Resident inspector, Prairie Island C. Brown Resident inspector, Big Rock Point G. Pirtle, Senior Security inspector -

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Approved by: J. W. McCormick Barger, Chief ,

Reactor Projects Branch 7

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EXECUTIVE SUMMARY Monticello Nuclear Generainig Station, Unit 1 NRC Inspection Report No. 50 203/97014(DRP)

This inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a 6 week period of resident inspection. This report also includes inspection follow up of a security-related issue by a regionalinspecto Qp_erations

  • Operator performance during routine operations and surveillance activities was good, and operations personnel conducted the planned reactor power decrease in a controlled manner. However, operations personnel were unaware that a condensate storage tank level instrumer;t was inoperable during a maintenance activity. (Section 01.1)

Maintenance

  • In general, the work associated with selected maintenance and sun aillance activities was conducted in a professional and thorough manner. For work on the residual heat removal service water system, excellent foreign material exclusion control as well as careful work accomplished in accordance with procedures were observed during the replacement of surge check valves. Good oversight by quality control and engineering personnel was also provided. (Section M1.1)

[ condensate storage tank),"was inadequate. The licensee initially failed to identify that operation with only one CST in service with the high pressure coolant injection (HPCI)

and reactor core isolation cooling (RCIC) pump suction valves lined up to the CST was in violatiori of Technical Gpecifications. In addition, the licensee aborted a routine surveillance test when the suppression pool supply valves re opened unexpectedly. An engineer did not fully evaluate plant conditions prior to requesting the initiation of the surveillance test. (Section M1.2)

  • The inspectc,rs identified that the surveillance tests for HPCI and RCIC steam line isolation valves could lead to preconditioning of the valves prior to determining the "as found" valve closure times. The system engineer initicted the appropriate corrective action document to resolve the issue prior to performance of the next test (Section M3.1)

Enoineerina a An engineer demonstrated a good questioning attitude and identified that continued operation with one CST and the HPCI and RC?C suction valves aligned to the #11 CST was contrary to Technical Specifications. (Section M1.0)

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  • Engineering personnel responded appropriately to an industry event concerning potential effects of ventilation en tank levels. (Section E2.1)
  • Engineering personnel demonstrated a good questioning attitude in identifying a vodex concem and a non conservative Technical Specification low level setpoint for the condensate storage tanks. The licensee initiated prompt corrective actions including establishing an administrative liinit and implementing procedure changes. (Section E2.2)

Plaat Support

  • During two security training drills, good communication was established between field and alarm station personnel, and the guard force responded appropriately to the " events,"

The licensee's critiques of the drills were informative and self critical. (Section S1)

  • The inspectors identified a violation conceming a vulnerability in the northwest portion of the vehicle barrier system which could allow a vehicle to gain unauthorized proximity to ,

vital equipment. (Section S8.1)

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Report Details Summary of Plant Status The Unit operated at or near 100 percent power during this period. On October 12,1997, operators reduced reactor power to 75 percent for (nat day to facilitate routine quar 1erly testin . Operations 01 Conduct of Operations 01.1 General Comments (71707)

Using Inspection Procedure 71707, the inspectors conducted frequent reviews of oregoing plant operations. The inspection included observations of control room evolutions, shift turnovers, and operator rounds. The inspectors also reviewed control room logbooks and operability determinations. Updated Safety Analysis Report (USAR) Section 13, " Plant Operations," was reviewed as part of the inspectio In general, the conduct of operations was acceptable. Operator performance during routine operations and surveillance activities was good. Discrepancies identified during surveillance testing were promptly communicated to operations management and resolved satisfactorily. Operators responded appropriately to annunciators and alarms, and conducted a planned reactor power decrease in a controlled manne However, operations personnel were unaware that a condensate storage tank level instrument was inoperable during a maintenance activity. This issue is discussed in Section M (

O2 Operational Status of Facilities and Equipment O2,1 Enaineered Safety Feature System Walkdowns !alpection Scope (71707)

The inspectors inspected the accessible areas of the Division 1 and 2,125 and 250-volt direct current (Vdc) batteries and the residual heat removal service water (RHRSW) systems. The purpose of the inspections was to assess the overall c!santiness of the areas and the material condition o' the equipment. Observations and Findinas Overall, the inspectors determined that the material condition of the batteries and RHRSW systems was acceptable. The electrolyte levels for the batteries were in the proper range and there was no evidence of leaks or corrosion on the individual cells, No operability concerns were identified, The inspectors made the following observations:

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hester was powered from circuit 14 of lighting panel L 36. Actually, thet circuit i powered the heater for the #14 RHRSW pump and the #13 RHRSW pump heater i was powered from lighting panel L 35, circuit 14, as shown on drawing -!

NL 36240 27, * Panel Scheduto L 35," Revision K. The system angineer stated s

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that the engineering staff was working on a comprehensive review and revision !

project for panel drawings and had already noted the discrepanc c

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  • The inspectors noted that the temperature in the battery rooms was above 80

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degrees and were concerned about the effect on battery life. The inspectors also noted the absence of a corrosion preventive compound on the battery fastener The electrical maintenance engineer stated that the elevated temperatures were 1 (

known and the possible effects on the batteries had been analyzed. The results of '

the battery performance tests did not indicate degradation or signs of premature

' failure. The engineer also stated that corrosion preventive compound was not required on the battery fasteners as the fasteners were made from stainless steeli :

The inspectors reviewed requirements and agreed with the engineer's statemen The inspectors reviewed USAR 8.5, "DC Power Supply Systems," and had no

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concerns between the systems as described and as found, Conclusions The inspectors found the Division 1 and 2 batteries and RHRSW systems to be in good material condition, The electrical maintenance engineer had maintained good records of >

. the batteries' condition and was knowledgeable of the current condition ,

ll. Maintenance M1 Conduct of Maintenance M General Comments

., Inspection Scope (62703 and 61726)

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The inspectors observed all or portions of selected maintenance and surveillance activities. Included in the inspection was a review of the surveillance procedures or work ,

orders (WOs) listed, as well as the appropriate USAR sections pertaining to these

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< Observations and Findinag

p" In general, the inspectors observed that the work associated with these activities was conducted in a professional and thorough manner. All work observed was performed with the work package present and in active use. Technicians were experienced and knowledgeable of their assigned tasks. The inspectors frequently observed supervisors and system engineers monitoring job progress, and quality control personnel were _ -

present whenever required by procedure. When applicable, appropriate radiation control

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measures were in plac .,

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l The following work was observed. Specific concerns or observations are provided where '

appropriate.

  • Surveillance Test 0255 07 IA 1, * Main Steam Valve Exercise Tests," Revision 10
  • Surveillance Test 10401, " Turbine Generator," Revision 30
  • Surveillance Test 0062, *RCIC [ reactor core isolation cooling] Steam Line High Area Temperature Test and Calibration Procedure, and Group 5 Isolation Valve Closure Test, Revision 11" The technical adequacy of this test is discussed in Section M3.1.
  • WO 9704242, Ocsign Change MP 96015013 " Install Cast Steel Surge Check Valve RHRSW 521" and WO 9704067, " Pre operational Test for RHRSW 521" T he inspectors reviewed the equipment isolation, work order, and design change procedures; observed the work; and discussed the design change with the system engineer. The inspectors observed excellent foreign material control and careful work accomplished in accordance with the procedures. Quality control verifications were properly conducted and the work was closely monitored by the system engineer. The inspectors also observed portions of the installation of similar cast steel surge check valves for the #12, #13, and #14 RHRSW pumps.
  • WO 9705257, " Preventive Maintenence on Low Pressure Coolant injection Valve MO 2015" The inspectors observed that the work was accomplished in an area posted as a neutron radiation area. Radiation work permit 187, Revision 0, specified for the job, did not require the maintenance workers to wear neutron dosimetry. The inspectors venfied that radiation personnel considered the actual neutron dose rate in the work area during the preparation of this radiation work permit. The licensee concluded that neutron dosimetry was not necessary. The inspectors had no further concerns.
  • Preventive Maintenance 4048, * Secondary Containment isolation Damper Maintenance," Revision 6 The inspectors noted that the procedure contained incorrect references to Technical Specifications (TSS). The system engineer initiated a procedure change to correct the errors.
  • Surveillance Teat 0058, "HPCI [high pressure coolant injection) Steam Line High Area Temperature Test and Calibration Procedure, and Group 4 Isolation Valve Closure Test," Revision 11
  • Surveillance Test 01981, "125 Vdc Battery Capacity Test," Revision 1
  • Preventive Maintenance 4525, "Nos.13 and 16 Battery Charger Preventive Maintenance," Revision 0

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  • Preventive Maintenance 4610, * Maintenance of On site Batteries and Battery Chargers at Monticello Nuclear Plant," Revision 10 Conclusions in general, these activities were conducted in a professional and thorough' manne Excel!ent foreign material exclusion control was observed during the replacement of surge check valves. Good oversight by quality control and engineering personnel was also provided.

M1.2 Problems Encountered Durina Condensate Storaae Tank (CST) Work Inspe_gtion Scope (62703 and 71707)

On September 8,1997, the operators isolated and drained the #12 CST to facilitate maintenance activities which included painting and preservation of the interior surfaces of the tank During the work, the licensee identified some problems including f ailure to meet TS requirements and inadequate changes to a routine surveillance. The inspectors reviewed the circumstances surrounding the findings and the licensee's corrective actions, Qbservations and Findinas Under normal operating conditions, the licensee cross-tied the #11 and #12 CSTs. The CSTs were the normal water supply for the initial operation of HPCI and RCIC turbin driven pumps. The suction valves for the HPCI and RCIC pumps automatically realigned to the suppression pool on low level (2 feet) in either CST using a one out-of two logic system of detectors on level columns attached to each tank. Prior to draining the #12 CST for maintenance work, the licensee jumpered the low level switch contacts for the

  1. 12 CST to prevent an automatic valve realignment on low CST water level. The licensee operated with only one CST lined up to the HPCI and RCIC suction pipin During the maintenance activities, the licensee identified the following problems:
  • On September 17,1997, while reviewing another issue with the CST level switches, a system engineer noted that TS Table 3.2.8, item C,a, required that both CST low level switches be operable or HPCI and RCIC suction valves were required to be lined up to the suppression pool. The TS also required the level channel be returned to an operable status within 30 days or the plant had to be placed in Startup, Refuel, or Shutdown mode. Additionally, TS 3.5.D.3 stated that the coretrols for the automatic transfer of RCIC pump suction may be inoperable for 30 days,if the pump suction was aligne,d to the suppression pool If this condition could not be met, TS 3.5.D.4 required an orderly reactor shutdown, The licensee identified that the current equipment configuration (the #12 CST level switch was bypassed and the HPCI and RCIC suction valves were aligned to the
  1. 11 CST) did not comply with TS Table 3.2.8 and TS 3.5.D 4. The operators immediately realigned the HPCI and RCIC suction valves to the suppression pool and removed the jumper. The licensee initiated a condition report to determine the causes for the TS oversight.- The failure to realign the HPCI and RCIC suction valves to the suppression pool after making one CST level instrument inoperable constituted a licensee-identified violation of TS Table 3.2.8 and is considered a

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non-cited violation, consistent with Section Vll of the NRC Enforcement Policy (NCV 50 263/97014-01(DRP)). The licensee's corrective actions to prevent i recurrence was documented in Licensee Event Report (LER) 50 263/9701 l

  • On September 22,1997, the licensee planned to perform surveillance test 0058,

HPCI Steam Line High Area Temperature Test and Calibration Procedure, arid Group 4 Isolation Valve Closure Test." The test as written assumed the HPCI suction valves were lined up to the CSTs however, due to plant conditions, the valves were lined up to the suppression pool. The system engineer realized that the group 4 isolation signal, generated during this test, would isolate the suppression pool water supply valves. The system engineer wrote a temporary change to the procedure to close those valves at the start of the test. During the test, the suppression poolwater supply valves re opened unexpectedly. The licensee determined that the system engineer failed te consider that with a low CST signallocked in from the drained #12 CST, the suppression pool valves would immediately reopen after an attempt to close them. The licensee stopped the test and performed it again two days later with another temporary change that also disabled the low CST level signa Conclusions An engineer demonstrated a gooa questioning attitude and identified that continued operation with one CST and the HPCI and RCIC suction valves aligned to the #11 CST was contrary to Technical Specifications. However, the planning and approval of Work Order 9501551, " Repair interior Coating of #12 CST," was inadequate. The licensee initially failed to identify that operation with only one CST in service and wRh the HPCI and RClO pump suction valves lined up to the CST was in violation of Technical Specifications, in addition, the licensee aborted a routine surveillance test when the suppression pool supply valves re opened unexpectedly. An engineer did not fully evaluate current plant conditions prior to requesting the performance of the surveillance test, M3 Maintanance Procedures and Documentation M3.1 Potential to Precondition Valves Before Measunpa Stroke Timej Inspection Scom (92902j The inspectors reviewed the following surveillance test procedures for containment isolation valves:

  • Surveillance Tust 0056, "HPCI High Steam Flow Sensor Test and Calibration Procedure, anj Group 4 Isolation Valve Closure Test," Revision 19

- Surveillance Test 0058,"HPCI Steam Line High Area Temperature Test and Calibration Procedure, and Group 4 Isolation Valve Closure Test," Revision 11

  • Surveillance Test 0060, RCIC High Steam Flow Sensor Test and Calibration Procedure, and Group 5 Isolation Valve Closure Test," Revision 19 B

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. Surveillance Test 0062, "RCIC Steam Une High Area Temperature Test and ,

Calibration Procedure, and Group 5 Isolation Valve Closure Test," Revision 11 l Qbservations and Findinas  ;

Technical Specification 4.7.D.1.a required the licensec to test power-operated,  !'

containment isolation valves for simulated automatic initiation and closure times once an operating cycle. The maximum ope: s6ng time for the valves was specified in USAR

, Table 5.2.3h. Fcr HPCI and RCIC steam lines, the licensee met this requirement by measuring the closure times for valves MO (motor operated) 2034 and MO 2035, the HPCI steam supply isolation valves, and MO 2075 and MO 2076, the RCIC steam supply isolation valves, during the performance of surveillance tests 0056 and 0060,-

respect!vely, in January of each yea The inspectors noted that surveillance test 0056 directed the operators to manually actuate the isolation relay which closed both MO 2034 and MO 2035, but only measured the closing time of MO 2034. The procedure then directed the operators to reset the -!

holation, reopen MO 2035, and then re. actuate the isolation relay to time MO 203 Thus, the measured closure time for MO 2035 might not have tspresented the as found condition of the velve because of preconditioning. The same situatioa existed in surveillance test 0060 for valves MO 2075 and MO 207 ,

in addition, the inspectors identified that re!ated tests 0058 and 0062 were often performed just before tests 0056 and 0060 During these tests, valves MO 2034,2035, 2075, and 2076 were cycled closed twice. The closure times were not measured in those tests. Thus, additiona! preconditioning might occur just prior tu measuring the cloture times to meet the TS requ!rement ,

The inspectors discussed the concern with the system engineer who subsequently issued condition report 97002542 with a resolution date prior to the next January performance of the surveillance tests. The inspectors also reviewed Information Notice 9716,

  • Preconditioning of Plant Structures, Systems, and Components Before ASME [American Society of Mechanical Engineers} Code Inservice Testing or Technical Specification Surveillance Testing," and noted that the licensee was evaluating the applicability of this Information Notic Conclusion $

The inspectors identified that the surveillance tests for HPCI and RCIC steam line

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isolation valves could lead to preconditioning of the valves prior to determining "as found" valve closure times.. The system engineer initiated the appropriate corrective action document to resolve the issue prior to performing the next tests.

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r Ill. Ennineerina E2 Engineering Support of Facilities and Equipment  :

E Enaineerina Personnel Review of Ventilation System Effects on Indicated CST Level .

i Inspection Scope (37551)

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NRC Information Notice 97 33 " Unanticipated Effect of Vemilation System on Tank Level Indications and Engineering Safety Features Actuation System Getpoint," discussed the effects of building pressure on tank levelindications During this inspection period, a :

system engineer evaluated the information Notice with respect to the CST level switche The inspectors reviewed the licensee's evaluatio Observations and Findinat The licensee determined that the condition discussed in Informat!on Notice 97 33 was applicable to the CST levelinstrumentation. The CST a were vented to the offgas .

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recombiner building which was assumed to be at atmnspheric plessure during accident conditions. However, the level columns and level switches on the CSTs were located in the r6 actor building which could be at less than almospheric pressure. The engineer concluded that the differential pressure betwecn the tank and the level detectors would cause the CST level to indicate higher than actuallevel. This higher indicated level would ,

delay the automatic realignment of HPCI and RCIC suction valves from the CSTs to the suppression pool on low CST level. This effect was also observed dunng a routine preventive maintenance activity which included securing the normal reactor bv.lding

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supply fans with the exhaust fans running. The licensee recorded reactor building l pressure and indicated CST level on a graph which clearly showed an inverse relationship between reactor building pressure and indicated CST level. The licensee ,

determined that indicated CST level decreased about one inch for each change in >

pressure of one inch water colum T-The licensee evaluated the as found setpoint of the level switches and determined that -

this phenomenon alone did not pose a concern since an adequate margin in level setpoints existed. The operability of the HPCI and RCIC systems was not affecte Conclusions

- Eng;neering personnel responded appropriately to an industry event concerning potential effects of ventilation on tank level E Non-conservative Setpoints for CST Level Instrumentation Due o Potential Voaexina Inspection Scope (71702)

On September 26,1997, the licensee identified that vortexing could introduce air into the

- HPCI and RCIC pumps prior to automatic realignment of the respective suction valves to l - the suppression pool. The inspectors assessed the licensee's actions to address this issue. The following documents were also reviewed:

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  • USAR Chapter 10.2.5, Revision 14, and Table 10.2-3, Revision 14
  • TS Table 3.2.0, Revision 143 l l
  • Safety Review Item 97 008, " Condensate Storage Tank Level Setpoint for HPCl/RCIC duction Transfer and Related USAR lssues,' Revision 0
  • Calculation 97-0232, " CST Suction Line Submergence for Vortex Concern,'

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  • Calculation 97 0235, *HPCI Suction Transfer from CST Setpoint Calculation,"

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  • NRC Regulatory Guide 1.82, " Water Sources for Long Term Recirculation Cooling Following a Loss of Coolant Accident,' Revision 1 b. Observations and Findinag Technical Specification Table 3.2.8 required that the low CST level setpoints be greater or equal to 24 inches above the bottom of the CST, The bases for this TS section documented an allowable rninus 6-inch deviation (18 inches from the bottom of the tank)

for as found data. During review into other CST levelissues discussed in this inspection report, on engineer identified that this TS requirement was non conservative with one CST available. The engineer determined that a higher suction inlet velocity resulted when HPCI and RCIC suction valves viere aligned to only one CST. The higher flow velocity affected vortexing and could introduce air in the suction piping of the HPCI and RCIC pumps, thus reducing system flow rates. The licensee made a 10 CFR 50.72 notification due to the potential of operating outside of the design basis when only one CST was on line between September 8 and 17,1997. The Gcensee later determined that vortexing could also occur with two CSTs in operation. Both HPCI and RCIC systems were aligaed to the suppression pool suction supply at the time of this discover The licensee performed calculations to determine a new setpoint to initiate HPCI and RCIC suction valve realignment. The inspectors noted that the calculations also considered building ventilation effects and instrument uncertainties. The licensee determined that for one and two CST operation, the level setpoint should be about 81 inches and 27 inches from the bottom of the tank, respectively. The inspectors had the following concerns and observations:

  • Methodoloav for determinina vortexina effects and pojential TS chanae: The licensee had difficulty determining en industry-accepted technique for calculating minimum submergence requireme.ds for the CST piping. The suction piping for the HPCI and RCIC pumps entered the CSTs horizontally and then turned down 90 degrees with the opening pointing down. However, as documented in Regulatory Guide 1.82, calculations for determining r,inimum submergence of suction piping assumed horizontal or vertical suction piping. The licensee initially determined submergence requirements assuming vertical piping and was evaluating other methods for determining vortexing effects. The licensee planned to resolve this methodology issue, recalculate a required low level setpoint, and submit a TS s,mendment request. The inspectors will review the cornpleted evaluation of the new setpoints as an Inspection Follow-up Item (IFl 50-263/97014 02(DRP)).
  • Administrative Limit Established: The licensee initiated work orders to recalibrate t

the levelinstruments with a new setpoint of 32 inches from the bottom of the tan ,

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No TS change was needed for this change since the TS allcwed the setpoint to be greater than 24 inches. The inspectors questioned which value the licensee would use to determine future as found operability since a deviation to 18 incnes j above the bottom of the tank was allowed in the TS bases. The licensee i

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responded that operability of HPCI and RCIC would be evaluated if a future as-found level setpoint was below 27 inches. The inspectors acknowledged that the licensee may modify these values based on the results of the final calculation l

SafetLSlanificance of Continued OAtration with One CST: The licensee identified that for operation with one available CST, the low level setpoint needed to be changed frorn 24 inches to about 82 inches from the bottom of the CST. The licensee determined that between August 31 and November 13,1995, cnd between September 8 and 17,1997, only one CST was aligned for HPCI and RCIC suction and the low CST level setpoint was set at about 24 inches. The licensee had not issued the LER associated with this event prior to the end of the inspection period. Following issuance of the LER, the inspectors will review ine licensee's assessment and determine the safety significance of this even Therefore, previous full reactor power operation with one available CST with a non-conservative level setpoint is an Untosolved item (URI 50 263/97014 03(DRP)).

  • Qp3.rftions Committee Members Approved Modif; cation for One CST Operation:

The Technical Specification required the two CST level channels be operable when the HPCI and RCIC suction valves were aligned to the CSTs. The licensee investigated methods which would meet the TS requirements and allow one CST to be unavailable. Engineering personnel designed a system modification to insta" a 5 foot spool piece on the existing levelinstrumentation piping. This would allow the licensee to change the level setpoint to about 82 inches to prevent vortexing. The licensee also planned to install additional piping to cross connect the levelinstrument from the out of service CST to the levelinstrument of the available CST. The licensee reasoned that this action would satisfy the Technical Specification requirement of two operable levelinstruments. The modification was approved by the operations committee; however, the modification was not implemented. The inspectors noted that CSTs were not required by TS and the licensee did not take credit for the CSTs in accident scenarios. However, it was l not clear whether TS Table 3.2.8 assumed one levelinstrument on each tank or l

would allow two levelinstruments on one tank. This issue is an IFl (50-263/97014 04(DRD)) pending clarification of the TS requirements and further review by the inspector c. Conclusions i

Engineering personnel demonstrated a good questioning in identifying the vortex concern and non-conservative TS setpoint. The licensee initiated prompt corrective actions including establishing an administrative limit and implementing procedure change i

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E8 Miscellaneous Engineering issues EG.1 (Closed) LER 50-263/M018. Revision 0: Average Power Range Monitors Scram Set Point Higher Than Allowed by the Updated Safety Analysis Caused by inadequate Engineering Revie This LER was previously discussed in inspection Report 50-263/92016(DRP),

Section 2.g(2). The inspectors verified through document review that the two long-term corrective actions in the LER, to revise the USAR end review other setpoint parameters used in the transient analysis, were completed.

E8.2 (Closed)1ER 50 26F9004. Revision 0: Electrical Storm Disables Plant Equipment on Two Occasion This LER was previously discussed in inspection Reports 50 263/94004(DRP),

Section 2.d, and 50 263/97002(DRP), Section E8.1. As discussed in the latter report, the one remaining corrective action was to modify spent fuel pool radiation monitor RM 17 453B to make it less susceptible to lightening induced failure. The licensee completed Work Order 9704331 on May 20,1997, to increase the size of the monitor's line fuse from 0.06 ampere to 0.15 ampere. The licensee documented in condition report 94001154 an extensive study of the histor/ of lightening induced failures at the plant and discussed other corrective actions in addition to those discussed in the LE [S, Plant Support R1 Conduct of Radiological Protection and Chemistry Controls (71750)

During rormal resider,t inspecion activities, routine observations were conducted in the area of radiation protection. The inspectors noted that radiation protection technicians provided support during maintenance and surveillance activities. On September 17, 1997, the inspectors noticed that an eL, iron dosimeter with an expired calibration date of March 9,1997, was placed in the in service dosimeter rack. The licensee immediately verified that the asdound condition of the dosimeter was within the calibration acceptance criteria and confirmed that no overexposures resulted from its previous use. Radia* ion protection procedure R.03.01, " instrument Requirements," Revision 13, requires instruments be removed from service if the calibration due date expired. The failure to follow procedure R.03.01 constituted a minor violation of TS 6.5.B. ,

(50-263/97014-05(DRP)) and is being treated as a Non-Cited Violation, consistent with Section IV of the NRC Enforctment Policy.

S1 Conduct of Security and Safeguards Activities (71750)

During normal resident inspection activities, routine observations were conducted in the areas of security and safeguards activities. No concerns were noted. On October 6, 1997, the inspectors observed two sseurity training dnlls. The inspectors noted that good communication was established between field and alarm station personnel and the guard force responded appropriately to the " events." The licensee's entiques of the drills were informative and self-critical. On October 7,1997, the inspectors met with the security

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i supervisor and other members of the security steff to discuss current security issue Topics included material condition of security-related equipment, personnel performance trending, and preparation activitif,s for the scheduled Operational Safeguards Response -  !

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Evaluatio i 58- Miscelianeous Security and safeguards issues 68,1 (Closed) URI (50-263/97005-01): The inspectors identified that from February 29,1996, to April 10,1997, the northwest portion of the vehicle barrier systern (VBS) was not  ;

i continuous and could allow a vehicle to gain unauthorized proximity to vital equipmen The licensee personnel had considered the initial design and installation of the VBS to be  :

adequate, and the vulnerability was not identified during routine observations and i inspections of the VBS. After further review, the inspectors concluded that the VBS *

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vulnerabikty was contrary to 10 CFR 73.55(c)(7) which requires the licensee to establish vehicle control measures to protect against the use of a land vehicle to gain unauthorized -

proximity to vital areas. Section 2C(3) of Amendment 58, dated December 13,1988, to Facility Operating License No. DPR 22 requires the licensee to maintain and fully l Implement all provisions of the security plan Section 20.1.1 of the facility security plan l states that a vehicle barrier system has been installed as required by and in accordance with applicable regulations The failure to follow 10 CFR 73.55 is a violation (VIO 50 263/97014 OS(DRS)) The inspectors reviewed the licensee's corrective actions  ;

including the immediate compensatory measures and installation of a permanent barrie The inspectors had no further concerns with this issu V. Manaaement Meetinas X1 Exit Meeting Summary  ;

On October 14,1997, the inspectors presented the inspection results to the plant manager and quality services manager. The licensee acknowledged the findings presented. The inspectors asked the licensee whether any materitals examined during the inspection should be considered proprietary, No proprietary information was ,

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i PARTIAL LIST or P!!RSONS CONTACTED . .;

Linertoet '

M.' Wadley, Vice President Nuclear Generation i M. Hammer; Plant Manager-  !

B. Day, Trainin; Manager  !

K. Jepson, Superintendent, Chemistry & Environmental Protection }

L. Nolan, General Superintendent safety Assessment i

. M. Onnen, General Superintendent Operations

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E. Reilly, General Superintendent Maintenance

- C. Schibonski, General Supe:Intendent Engineering [

A. Ward, Manager Quality Services l L. Wilkerson, SuperinteMont Security .i

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- J. Windschill, General Superintendant, Radiation Protection

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i INSPECTION PROCEDURES USED

IP 37551: Onsite Engineering  !

IP 61726: Gurveillance Observations l IP B2703: Maintenance Observations  :

IP 71707: Plant operations . l lP 71750: Plant Support  !

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ITEMS OPENED, CLOSED, AND DISCUSSED

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4 50 263/07014 02(DRP) IFl The completion of the licensee's evaluation to detennine new TS setpoints for low CST level instrumentation l

50-263/97014 03(DRP) URI Determination of the safety significance of previous full j reactor power operation with one available CST with a non-  ;

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conservative level setpoint 50 263/t!014 04(DRP) IFl Clarification c! t TS requirements for CST level instrumentatiwi Closed -

50 263/97005 01(DRP) URI Potential for vehicle barrier system effectiveness being

circumvented 50-2S3/97014 01(DRP) NCV Failure to realign the HPCI and RCIC suction valves to the {

suppression pool after making one CST level channel inoperable 50 263/07014 05(DRP) NCV Failure to follow radiation protection; dosimeter out of-calibration 50-263/97014 06(DRS) \tlO Vehicle barrier system coald allow unauthorized proximity to

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vital equipment

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50 263/92018 LER Averago power range monitors scram setpoint higher than allowed by the updated safety analysis report caused by inadequate engiraeering review  :

50 263/94004 LER- Electrical storm disab!cs plant equipment on two occasions i

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Ll8T OF ACRONYMS USED i

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ASME Amencan Society of Mechanical Engineers i CFR Code of Federal Regulations l

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CST _ Condensate Storage Tank t CRP Division of Reactor Projects  !

-f)RS Division of Reactor Safety l HPCI High Pressure Coolant injection = l

- IFl Inspection Follow-up item LER Licensee Event Report Mo Motor Operatou-NCV Non Cited Violation .

NRC Nuclear Regulatory Commission l

. NS Northern States Power  ;

' RCIC Reactor Core Isolation Cooling RHRSW Residual Heat Removal Service Water i TS Technical Specification i URI Unresolved item _ >

USAR Updated Safety Analysis Report  :

.VBS Vehicle Barrier System l Vdc Volt Direct Current WO Work Order -;

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